1. Technical Field
The present invention relates to a method of separating uranium from at least fission products in irradiated nuclear fuel.
2. Related Art
For many years the preferred method of reprocessing irradiated nuclear fuel has been the established Purex process. The Purex process includes an initial stage or ‘head-end’ in which the fuel rods first are cut into shorter lengths and then are exposed to hot nitric acid which leaches out the irradiated nuclear fuel from inside the ZIRCALOY (tradename) cladding. ZIRCALOY is a trademark designating a family of zirconium alloys containing tin and zirconium. The unreacted ZIRCALOY cladding is collected and disposed of as Medium Activity (MA) waste. After the ‘head-end’ is an extraction stage, in which the nitric acid solution which contains the uranium and plutonium as well as fission products is subjected to a solvent extraction cycle to separate the uranium and plutonium from the fission products. In subsequent stages the uranium and plutonium are separated and purified.
It remains desirable to improve aspects of the Purex process. For example, it would be advantageous to simplify the cutting and dissolution steps in the ‘head-end’. It would also be desirable to reduce the volume of acid and solvent used and thereby the volume of waste generated.
The present invention aims to overcome these problems by way of an alternative method to the Purex process.
A reprocessing method has been disclosed some time ago in U.S. Pat. No. 3,012,849 and U.S. Pat. No. 3,145,078 comprising converting the uranium and zirconium in the irradiated fuel and cladding respectively to uranium and zirconium fluoride complexes and then separating the uranium complex. In that method the fluoride complexes are formed by reaction of the fuel and cladding with a mixture of HF and either NOF or metal fluoride (denoted MF). However, when NOF is used, large volumes of NOx is produced and when MF is used large volumes of solid waste are generated. Also, the method could be improved in terms of overall separation efficiency.
Previously, in Synth. React. Inorg. Met-Org. Chem. 26 (1996) 139, ammonium bifluoride has been proposed to treat ZIRCALOY to convert it to a convenient storage form as part of a nuclear waste storage scheme. Also previously, in Chem. Eng. Prog. 50 (1954) 230, ammonium bifluoride has been used in a process for fabricating uranium metal nuclear fuel.
DD 301,016 discloses use of aqueous solutions of HF and/or ammonium fluoride and nitric acid for etching to achieving a corrosion resistant layer on the surface of fuel elements. U.S. Pat. No. 3.832,439 discloses use of an aqueous solution of ammonium fluoride for dissolving zirconium cladding.